Abstract Predicting the dramatic changes in mechanical and physical properties caused by irradiation damage is key for the design of future nuclear fission and fusion reactors. Self-ion irradiation provides an attractive tool for mimicking the effects of neutron irradiation. However, the damaged layer of self-ion implanted samples is only a few microns thick, making it difficult to estimate macroscopic properties. Here we address this challenge using a combination of experimental and modelling techniques. We concentrate on self-ion-implanted tungsten, the front-runner for fusion reactor armour components and a prototypical bcc material. To capture dose-dependent evolution of properties, we experimentally characterise samples with damage levels from 0.01 to 1 dpa. Spherical nano-indentation of grains shows hardness increasing up to a dose of 0.032 dpa, beyond which it saturates. Atomic force microscopy (AFM) measurements show pile-up increasing up to the same dose, beyond which large pile-up and slip-steps are seen. Based on these observations we develop a simple crystal plasticity finite element (CPFE) model for the irradiated material. It captures irradiation-induced hardening followed by strain-softening through the interaction of irradiation-induced-defects and gliding dislocations. The shear resistance of irradiation-induced-defects is physically-based, estimated from transmission electron microscopy (TEM) observations of similarly irradiated samples. Nano-indentation of pristine tungsten and implanted tungsten of doses 0.01, 0.1, 0.32 and 1 dpa is simulated. Only two model parameters are fitted to the experimental results of the 0.01 dpa sample and are kept unchanged for all other doses. The peak indentation load, indent surface profiles and damage saturation predicted by the CPFE model closely match our experimental observations. Predicted lattice distortions and dislocation distributions around indents agree well with corresponding measurements from high-resolution electron backscatter diffraction (HR-EBSD). Finally, the CPFE model is used to predict the macroscopic stress-strain response of similarly irradiated bulk tungsten material. This macroscopic information is the key input required for design of fusion armour components.

中文翻译：

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自离子辐照钨的改性变形行为：结合纳米压痕、HR-EBSD 和晶体塑性研究

摘要 预测由辐照损伤引起的机械和物理特性的剧烈变化是未来核裂变和聚变反应堆设计的关键。自离子辐射为模拟中子辐射的影响提供了一种有吸引力的工具。然而，自离子注入样品的损伤层只有几微米厚，很难估计宏观特性。在这里，我们结合使用实验和建模技术来解决这一挑战。我们专注于自离子注入钨、聚变反应堆装甲部件的领跑者和原型 bcc 材料。为了捕捉特性的剂量依赖性演变，我们通过实验表征具有 0.01 到 1 dpa 损伤水平的样品。颗粒的球形纳米压痕显示硬度增加到 0 剂量。032 dpa，超过它就会饱和。原子力显微镜 (AFM) 测量显示堆积增加到相同剂量，超过此剂量可以看到大堆积和滑动步骤。基于这些观察，我们开发了一种用于辐照材料的简单晶体塑性有限元 (CPFE) 模型。它通过辐照诱导缺陷和滑动位错的相互作用捕获辐照诱导硬化，然后是应变软化。辐照诱导缺陷的抗剪切性是基于物理的，通过对类似辐照样品的透射电子显微镜 (TEM) 观察进行估计。模拟了剂量为 0.01、0.1、0.32 和 1 dpa 的原始钨和植入钨的纳米压痕。0 的实验结果只拟合了两个模型参数。01 dpa 样品，所有其他剂量保持不变。CPFE 模型预测的峰值压痕载荷、压痕表面轮廓和损伤饱和度与我们的实验观察结果非常吻合。压痕周围的预测晶格畸变和位错分布与高分辨率电子背散射衍射 (HR-EBSD) 的相应测量结果非常吻合。最后，CPFE 模型用于预测类似辐照的块状钨材料的宏观应力应变响应。这种宏观信息是设计融合装甲部件所需的关键输入。压痕周围的预测晶格畸变和位错分布与高分辨率电子背散射衍射 (HR-EBSD) 的相应测量结果非常吻合。最后，CPFE 模型用于预测类似辐照的块状钨材料的宏观应力应变响应。这种宏观信息是设计融合装甲部件所需的关键输入。压痕周围的预测晶格畸变和位错分布与高分辨率电子背散射衍射 (HR-EBSD) 的相应测量结果非常吻合。最后，CPFE 模型用于预测类似辐照的块状钨材料的宏观应力应变响应。这种宏观信息是设计融合装甲部件所需的关键输入。